Reprocessing is commonly used to recover plutonium, uranium, and other useful materials from spent nuclear fuel. The current standard method for reprocessing is the PUREX method, which is a liquid-liquid extraction method that can extract both uranium and plutonium independently of each other and from other fission products. Reprocessing methods generate liquid waste that includes both high activity waste (also referred to a high level waste), which carries many of the fission products and transuranic elements generated in the core, and low activity waste (also referred to a low level waste), which carries low activity fission products, actinides, and a plurality of different salts. The low activity waste is further treated to remove actinides and fission products to produce a decontaminated salt solution that, while exhibiting low activity levels, must be properly treated and stored to prevent release of contaminants.
Treatment of the decontaminated salt solution includes mixing it with a blend of cementitious materials to form a grout mixture. Upon curing of the grout mixture through hydration reactions a hardened monolithic cementitious waste form known as saltstone is formed. This particular treatment process is carried out at the Savannah River Site nuclear reservation in South Carolina, USA.
Some contaminants contained in such waste forms (e.g. chromium and technetium in saltstone, which are precipitated as very insoluble compounds under alkaline reducing conditions and are soluble under alkaline oxidizing conditions) exhibit a variable solubility depending upon their oxidation state, with the reduced form of the contaminants showing lower solubility. As such, reducing conditions are created in the waste form to slow and/or prevent release of contaminants. For instance, blast furnace slag is utilized in combination with a calcium silicate-based cement (e.g., Portland cement) in forming the saltstone waste form. Blast furnace slag significantly lowers the Eh, or redox potential, relative to traditional cements and thus serves to increase reducing conditions of the saltstone.
Desirably, the reduction capacity of the saltstone will persist over an extended performance period (e.g., 10,000 years) with rates of contaminant release dominated by slow changes in physiochemical properties. Even so, over time the chemical properties of the saltstone will vary as the result of exposure to air, groundwater, and other environmental factors. As a result the pH will decrease and the Eh potential will increase, as the saltstone oxidizes. For example, the rate of oxidation of the cement-slag based waste form, saltstone, has been calculated to be less than 0.5 millimeters per year based on oxygen diffusion models where diffusion rates change as a function of the square root of time. Models have considered liquid phase transport and a diffusion dominated process independent of flow through a fractured network and a shrinking core model.
Unfortunately, models are theoretical in nature, and require experimental data for verification. What are needed in the art are testing methods that can determine the rate of oxidation of cured cementitious materials. More specifically, what are needed are methods that can trace the location of an oxidation front in cementitious materials. Such methods would be of great use to verify modeling assumptions for long-term performance of materials such as saltstone, for long-term storage and disposal unit design, as well as to provide a basis for potential processing changes, such as clean cap installation criteria based on maximum allowable saltstone atmospheric exposure time.